L-MT-03-039, Revised Monticello Nuclear Generating Plant Technical Specifications Bases Pages

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Revised Monticello Nuclear Generating Plant Technical Specifications Bases Pages
ML031540681
Person / Time
Site: Monticello Xcel Energy icon.png
Issue date: 05/22/2003
From: Denise Wilson
Nuclear Management Co
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
L-MT-03-039
Download: ML031540681 (25)
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7 NM CO Monticello Nuclear Generating Plant Committed to Nuclear Exce Operated by Nuclear Management Company, LLC May 22, 2003 L-MT-03-039 Technical Specification 6.8.K US Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555 MONTICELLO NUCLEAR GENERATING PLANT DOCKET NO. 50-263 LICENSE NO. DPR-22 TECHNICAL SPECIFICATION BASES PAGES Using the Monticello Technical Specification Bases Control Program, Monticello Technical Specification Bases pages have been changed. The affected pages are designated with the amendment applicable at the time and the suffix "a." The changes are summarized in Attachment 1. Marked up pages applicable at the time the changes were made are provided in Attachment 2. A final typed copy of the changed pages that are applicable, for entry into the NRC authority copy, are provided in Attachment 3. The current copy of our list of effective pages and record of revision is attached for your information, as Attachment 4.

Please contact John Fields at 763-295-1663 with any questions.

David L. Wilson Vice President Monticello Nuclear Generating Plant - Summary of Technical Specification Bases Changes (TSBC) - Monticello Technical Specification Bases Pages Marked Up With Changes - Revised Monticello Technical Specification Bases Pages - Monticello Technical Specification List of Effective Pages and Record of Revision cc:

Regional Administrator-lIl, NRC NRR Project Manager, NRC Resident Inspector, NRC Minnesota Department of Commerce 2807 West County Road 75

  • Monticello, Minnesota 55362-9637 Telephone: 763.295.5151
  • Fax: 763.295.1454 C)

ATTACHMENT I NUCLEAR MANAGEMENT COMPANY, LLC MONTICELLO NUCLEAR GENERATING PLANT DOCKET 50-263 May 20, 2003

SUMMARY

OF TECHNICAL SPECIFICATION BASES CHANGE (TSBC) 1 page follows 1-1

ATTACHMENT I Summary of Technical Specification Bases Change (TSBC)

Following is a summary of the bases changes forwarded herein. The changes have been processed in accordance with the Monticello Technical Specification Bases Control Program described in Technical Specification 6.8.K.

TSBC-133a Technical Specification Involved - 4.5.A.3 & 4.5.D.1 Pages affected - 111 & 114 Summary of Change: Modification of statement regarding adequate steam flow for the High Pressure Coolant Injection and Reactor Core Isolation Cooling system low and high pressure tests. The statement for each system test now contains a minimum turbine bypass valve open position. The Bases description was changed to be consistent with NUREG-1433, Rev. 2 language.

TSBC-1 35a Technical Specification Involved - 3.2, Table 3.2.1 Page affected - 64 Summary of Change: This TSBC adds language to the Technical Specification Bases that clarifies that the head cooling function, although abandoned, still performs a containment isolation function for penetration X-17.

1-2

ATTACHMENT 2 NUCLEAR MANAGEMENT COMPANY, LLC MONTICELLO NUCLEAR GENERATING PLANT DOCKET 50-263 May 20, 2003 MONTICELLO TECHNICAL SPECIFICATION BASES PAGES MARKED UP WITH CHANGES 4 pages follow 2-1

ATTACHMENT 2 Monticello Technical Specification Bases Pages Marked Up With Changes This attachment consists of Monticello Technical Specification bases page marked up with changes. The pages included are listed below:

Pae 64 111 114 2-2

Bases 3.2:

In addition to reactor protection instrumentation which initiates a reactor scram, protective instrurrientation has been provided which initiates action to mitigate the consequences of accidents which are beyond the operators ability to control, or terminate a single operator error before it results in serious consequences. This set of specifications provides the limiting conditions of operation for the primary system isolation function, initiation of the emergency core cooling system, and other safety related functions. The objectives of the Specifications are (i) to assure the effectiveness of the protective instrumentation when required1 and (ii) to prescribe the trip settings required to assure adequate performance. This set of Specifications also provides the limiting conditions' of operation for the control rod block system.

Isolation valves are installed in those lines that penetrate the primary containment and must be isolated during a loss of coolant accident so that the radiation dose limits are not exceeded during an accident condition. Actuation of these valves is initiated by.

protective instrumentation shown in Table 3.2.1 which senses the conditions for which isolation is required. Such instrumentation must be available whenever primary containment integrity is required. The objective is to isolate the primary containment sothat the guidelines of 1) CFR 100 are not exceeded during an accident.

The instrumentation which initiates primary system isolation is connected in a dual bus arrangement. Thus, the discussion given in the bases for Specification 3.1 is applicable here.

The low reactor water level instrumentation is set to trip when reactor water level is > 7' on the instrument. This corresponds to a lower water level inside the shroud'at 100% power due to the pressure drop across the dryer/separator. This has been accounted for in the affected transient analysis. This trip initiates closure of Group 2 primary containment isolation valves. Reference Section 7.7.2.2 FSAR. The trip-setting provides assurance that the valves will be closed before perforation of the clad occurs even for the maximum break in that line and therefore the setting is adequate. The head cooling valves no longer function to provide head cooling, but continue to provide containment isolation for penetration X-1 7.

The low low reactor water level instrumentation is set to trip when reactor water level is > -48". This trip initiates closure of the Group 1 and Group 3 Primary containment isolation valves, Reference Section 7.7.2.2 FSAR, and also activates the ECC systems and starts the emergency diesel generators.

3.2 BASES

-64 04/24/03 Amendment No. 65, 81, Opa, 102,117,128, 135a I

Bases 3.5/4.5 (Continued):

The surveillance requirements provide adequate assurance that the LPCI system will be operable when required. Although all active components are testable and full flow can be demonstrated by recirculation through a test loop during reactor operation, a complete functional test requires reactor shutdown. 'The pump discharge piping is maintained full to prevent water hammer damage to piping and'to start cooling at the earliest moment.

'The high pressure coolant injection (HPCI) system is provided-to assure that the reactor core is adequately cooled to limit fuel clad temperature in the event of a small break in the reactor coolant system and loss of'coolant which does not result in rapid depressurization of the reactor vessel. The HPCI system permits the reactor to be shut down while maintaining sufficient reactor vessel water level inventory,until the vessel is depressurized. The HPCI system continues to operate until reactor vessel pressure is below the pressure at which Core Spray system operation or LPCI mode of the RHR system operation maintains core cooling.

The flow tests for the HPCI System are performed at two different pressure ranges such that the system capability to provide rated flow is tested at both the higher and lower operating ranges of the system. Additionally, adequate steam flow must be passing through the main turbine or turbine bypass valves to continue to control reactor pressure when the HPCI System diverts steam flow. Reactor steam pressure must be > 950 psig to perform SR 4.5.A.3.a and 165 psig to perform SR 4.5.A.3.b. Adequate steam flow is represented by Xat least 0.8 turbine bypass valves open total steam flow 1 6 lb/hr. Reactor startup, and pressure increase'to c 165 psig, is allowed prior to performing the low pressure surveillance test because the reactor pressure is low and the time allowed to satisfactorily perform the surveillance test is short. Therefore, pressure may be raised above 150 psig, but' 165 psig to perform this surveillance without entering an LCO for the HPCI System. The reactor pressure is allowed to be increased to normal operating pressure once the low pressure test has been satisfactorily completed since there would be no indication or reason to believe that HPCI is inoperable.

Sufficient time is needed after adequate pressure and flow are achieved to perform these tests. Therefore, SR 4.5.A.3.a and SR 4.5.A.3.b are modified by a note which states that the surveillances are not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after the reactor steam pressure and flow are adequate to perform the test.

With the HPCI system inoperable, adequate core cooling is assured by the operability of the redundant and diversified automatic depressurization system and both the Core Spray and LPCI systems. In addition, the reactor core isolation cooling (RCIC) system, a system for which no credit is taken in the safety analysis, will automatically provide makeup, at reactor operating pressures on a reactor low waterlevel condition. The HPCI out-of-service period of 14 days is based on the verified operability of the RCIC system and the redundant and diversified low pressure core cooling systems. Verification of RCIC operability may be performed as an administrative check by examining logs or other information to determine if RCIC is out of service for maintenance or other reasons. It does not mean to perform' the surveillance needed to demonstrate the operability'of the RCIC system.

3.5/4;5 BASES 111 03/28/03 Amendment No. 37, 63, 77, 79, OOa, 122, 133a

Bases 3.5/4.5 (Continued):

at least 0.8 turbine bypass'valves o p

e nh~total ctoa flowŽ10p 6 bh. Reactor startup, and pressure increase to c 165 psig, is allowed prior to performing the low pressure surveillance test because the reactor pressure is low and the time'allowed to satisfactorily perform the surveillance test is short. Therefore, pressure may be raised above 150 psig, but < 165 psig to perform this surveillance without entering an LCO for the RCIC System. The reactor pressure is allowed to be increased to normal operating pressure once the low pressure test has been satisfactorily completed since there would be no indication or reason to believe that RCIC is inoperable.

Sufficient time is needed after adequate' pressure and flow are achieved to perform these tests. Therefore, SR 4.5.D.1.a and SR 4.5.D.1.b.

are modified by a note which states that the surveillances are not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after the reactor steam'pressure and flow are adequate to perform the test.

The surveillance requirements provide adequate assurance that the RCIC system will be operable when required. All active components are testable and full flow can be demonstrated by recirculation through a test loop during reactor operation. The pump discharge piping is.

maintained full to prevent water hammer damage and to providecooling at the earliest moment.

E.

Cold Shutdown and Refueling Requirements The purpose of Specification 3.5.E is to assure that sufficient core cooling equipment is available at all times. It is during refueling outages that major maintenance is performed and during such time that all core and containment'spray/cooling subsystems may be out of service.

This specification allows all core and containment spray/cooling subsystems to be inoperable provided no work is being done which has the potential for draining the reactor vessel. Thus events requiring core cooling are precluded.

Specification 3.5.E.2 recognizes that concurrent with control rod drive maintenance during the refueling outage, it may be necessary to drain the suppression chamber for maintenance or for the inspection required by'Specification 4.7.A.1. In this situation, a sufficient inventory of water is maintained to assure adequate core cooling in the unlikely event of loss of control rod drive housing or instrument.:

thimble seal integrity.

3.5/4.5 BASES 114 03/28/03 Amendment No. 21, 17, 70, 97, 1Oa, 102, 122, 133a

ATTACHMENT 3 NUCLEAR MANAGEMENT COMPANY, LLC MONTICELLO NUCLEAR GENERATING PLANT DOCKET 50-263 May 20, 2003 REVISED MONTICELLO TECHNICAL SPECIFICATION BASES PAGES 4 pages follow 3-1

ATTACHMENT 3 Revised Monticello Technical Specification Bases Pages This attachment consists of the revised Monticello Technical Specification Bases pages that incorporate the change. These pages should be entered into the NRC Authority copies of Technical Specifications. The pages included are listed below:

Paqe 64 111 114 3-2

Bases 3.2:

In addition to reactor protection instrumentation which initiates a reactor scram, protective instrumentation has been provided which initiates action to mitigate the consequences of accidents which are beyond the operators ability to control, or terminate a single operator error before it results in serious consequences. This set of specifications provides the limiting conditions of operation for the primary system isolation function, initiation of the emergency core cooling system, and other safety related functions. The objectives of the Specifications are (i) to assure the effectiveness of the protective instrumentation when required, and (ii) to prescribe the trip settings required to assure adequate performance. This set of Specifications also provides the limiting conditions of operation for the control rod block system.

Isolation valves are installed in those lines that penetrate the primary containment and must be isolated during a loss of coolant accident so that the radiation dose limits are not exceeded during an accident condition. Actuation of these valves is initiated by protective instrumentation shown in Table 3.2.1 which senses the conditions for which isolation is required. Such instrumentation must be available whenever primary containment integrity is required. The objective is to isolate the primary containment so that the guidelines of 10 CFR 100 are not exceeded during an accident.

The instrumentation which initiates primary system isolation is connected in a dual bus arrangement. Thus, the discussion given in the bases for Specification 3.1 is applicable here.

The low reactor water level instrumentation is set to trip when reactor water level is >7" on the instrurment. This corresponds to a lower water level inside the shroud at 100% power due to the pressure drop across the dryer/separator. This has been accounted for in the affected transient analysis. This trip initiates closure of Group 2 primary containment isolation valves. Reference Section 7.7.2.2 FSAR. The trip setting provides assurance that the valves will be closed before perforation of the clad occurs even for the maximum break in that line and therefore the setting is adequate. The head cooling valves no longer function to provide head cooling, but continue to provide containment isolation for penetration X-1 7.

The low low reactor water level instrumentation is set to trip when reactor water level is -48". This trip initiates closure of the Group 1 and Group 3 Primary containment isolation valves, Reference Section 7.7.2'2 FSAR, and also activates the ECC systems and starts the emergency diesel generators.

3.2 BASES 64 04/2 4/03 Amendment No. 65 81--1O0,02,'117, 126, 135a

Bases 35/4.5 (Continued):

The surveillance requirements provide adequate assurance that the LPCI system will be operable when required. Although all active components are testable and full flow can be demonstrated by rcirculation through a test loop during reactor operation, a complete functional test requires reactor shutdown. The pump discharge piping is maintained full to prevent water hammer damage to piping and to start cooling at the earliest moment.

The high pressure coolant injection (HPCI) system is provided to assure that the reactor core is adequately cooled to limit fuel clad temperature in the event of a small break in the reactor coolant system and loss of coolant which does not result in rapid depressurization of the reactor vessel. The HPCI system permits the reactor to be shut down while maintaining sufficient reactor vessel water level inventory until the vessel is depressurized. The HPCI system continues to operate until reactor vessel pressure is below the pressure at which Core Spray system operation or LPCI mode of the RHR system operation maintains core cooling.

The flow tests for the HPCI System are performed at two different pressure ranges such that the system capability to provide rated flow is tested at both the higher and lower operating ranges of the system. Additionally, adequate steam flow must be passing through the main turbine or turbine bypass valves to continue to control reactor pressure when the HPCI System diverts steam flow. Reactor steam pressure must be Ž950 psig to perform SR 4.5.A.3.a and 165 psig to perform SR 4.5.A.3.b. Adequate steam flow is represented by at least 0.8 turbine bypass valves open. Reactor startup, and pressure increase to <165 psig, is allowed prior to performing the low.

pressure surveillance test because the reactor pressure is low and the time allowed to.satisfactorily perform the surveillance test is short.

Therefore, pressure may be raised above 150 psig, but c 165 psig to perform this surveillance without entering an LCO for the HPCI System. The reactor pressure is allowed to.be increased to normal operating pressure once the low pressure test has been satisfactorily completed since there would be no indication or reason to believe that HPCI is inoperable.

Sufficient time is needed after adequate pressure and flow are achieved to perform these tests. Therefore, SR 4.5.A.3.a and SR 4.5.A.3.b are modified by a note which states that the surveillances are not required to bJe performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after the reactor steam pressure and flow are adequate to perform the test.

With the HPCI system inoperable, adequate core cooling is assured by the operability of the redundant and diversified automatic depressurization system and both the Core Spray and LPCI systems. In addition, the reactor core isolation cooling (RCIC) system, a system for which no credit is taken in the safety analysis, will automatically provide makeup at reactor operating pressures on a reactor low water level condition. The HPCI out-of-service period, of 14 days is based on the verified operability of the RCIC system and the redundant and diversified low pressure core cooling systems. Verification of RCIC operability may be performed as an administrative check by examining logs or other information to determine if RCIC is out of service for maintenance or other reasons. It. does not mean to perform the surveillance needed to demonstrate the operability of the RCIC system.

3.5/4.5 BASES 111 03/28/03 Amendment No. 37, 63, 77, 79, 00a, 122, 133a

Bases 3.5/4.5 (Continued):

at least 0.8 turbine bypass valves open. Reactor startup, and pressure increase to < 165 psig, is allowed prior to performing the low pressure surveillance test because the reactor pressure is low and the time allowed to satisfactorily perform the surveillance test is short.

Therefore, pressure may be raised'above 150 psig, but *165 psig to perform this surveillance without entering an LCO for the RCIC System. The reactor pressure is allowed to be increased to normal operating pressure once the low pressure test has been satisfactorily completed since there would be no indication or reason to believe that RCIC is inoperable.

Sufficient time is needed after adequate pressure and flow are achieved to perform these tests. Therefore, SR 4.5.D.1.a and SR 4.5.D.1.b are modified by a note which states that the surveillances are not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after the reactor steam pressure and flow are adequate to perform the test.

The surveillance requirements provide adequate assurance that the RCIC system-will be operable when required. All active components are testable and full flow can be demonstrated by recirculation through a test loop during reactor operation. The pump discharge piping is maintained full to prevent water hammer damage and to provide cooling at the earliest moment..

E.

Cold Shutdown and Refueling Requirements The purpose of Specification 3.5.E is to assure that sufficient core cooling equipment is available at all times. It is during refueling outages that major maintenance is performed and during such time that all core and containment spray/cooling subsystems may be out of service.

This specification allows all core and containment spray/cooling subsystems to be inoperable provided no work is being done which has the potential for draining the reactor vessel. Thus events requiring core cooling are precluded.

Specification 3.5.E.2 recognizes that concurrent with control rod drive maintenance during the refueling outage, it may be necessary to drain the suppression chamber for maintenance or for the inspection required by Specification 4.7.A.1. In this situation, a sufficient inventory of water is maintained to assure adequate core cooling in the unlikely event of loss of control rod drive housing or instrument thimble seal integrity..

3.5/4.5 BASES 114 03/28/03 Aniendment No. 21, 17, 7, 97,- 100a, 1 02,1-22T 133a

ATTACHMENT 4 NUCLEAR MANAGEMENT COMPANY, LLC MONTICELLO NUCLEAR GENERATING PLANT DOCKET 50-263 May 20, 2003 MONTICELLO TECHNICAL SPECIFICATION LIST OF EFFECTIVE PAGES AND RECORD OF REVISION 11 pages follow 4-1

ATTACHMENT 4 Monticello Technical Specification List of Effective Pages and Record of Revision This attachment consists of the current Monticello Technical Specification List of Effective Pages and Record of Revision. The pages included are listed below:

Paqe A

B C

D E

F G

H 4-2

MONTICELLO NUCLEAR GENERATING PLANT APPENDIX A TECHNICAL SPECIFICATIONS RECORD OF REVISIONS Amend Paae No.

A 135a B

134 C

115 D

115 E

115 F

115 G

115 H-119 I

130 J

135a i -

128 ii 104 iii 120 iv 128 v

120 vi 121 vii 122 1

119 2

70 3

21 4

102 5

120 5a 120 6

128 7

128 8

128 9

128 10 128 11 128 12 128 25a 127 25b 127 25c 127 25d 127 26

.5

27.

81 27a 81 28 128 29 128 30 103 31

.104 32 103 33 103 34 83 35 00a Amend Page No.

36 128 37 128 38 128 39 129b 40 129b 42 103 45 0

46 70 46a 37 47 40, 48 89 49 128

-50.

128 50a 117 51 117 51a 117 52 128 53 128 54 128 55 103 56 102 57 70 58 84.

58a 29 59 128 59a 103 60 128 60a 31 60b 62 60c 30 60d 128 60e 89 61

. 104 62 117 63 117 63a 117 64.

135a 65 117

  • 66 119a 67 117 68 129b 69 129b 69a 129b 70 117 71 100a Amend Page No.

71 a 72 76 77

  • 78 79 80.

81 82 82a 83 83a 84 85 86 87 88 89 90 91 92 93 94 95 96 97 98 99 100 101 102 103 104 105 106 107 108 109 110 111 112 113 114 115 121 129b 104 0

86 0

0 29.

3 123 63 24 24 1 00a 1 00a 100a 1 00a 1 00a 104 100a 123 1 00a 122 106 77 77 57 56

-104 100a 122 122 122 122 122 79 97 128 100a 1 00a 133a 130a 130a 133a 13Qa 0-Armend Paae No..

122 135 123 117 124 121 125 104 126 104 126a 87 127 128 128 42 129 122 130 82.

131 122 132 39 132a 122 133 106 134 133 135 133 136 133 137 0

138 100a 145 118a 146 135 147 107 148 117 149 1 OQa 150 128 151 128 153 100a 154 129a-155 122 156 93 157 130 158 132 159 132 160 132 163 130 164 104 165 130 166 130 167 112 168 94-169 94 170 130 171 130

.171a 130 172 71 A

Amendment No. 135a 04/24/03

MONTICELLO NUCLEAR GENERATING PLANT APPENDIX A TECHNICAL SPECIFICATIONS RECORD OF REVISIONS Amend Paae No.

175

.107 175a -

117 176 100a 177 130 178 100a 179 123a 180 128a 181 130 182 130 182a 130 183 117 184 100a 185 134 188 104 189 130 190 130 191 0

192 121 193 121 196 126a 197 121 198 121 199 51 200 129 201 129 202 129 203 41 204 129 204a 129 205 129 206 0

207 123 208 63 209 123 209a 100a 210 100a 211 131 212 109 213 99 216 100a 217 128 218 120 223 119 224 119 Amend Page No.

225 119 226 119 229a.

63 229b 104 229c 104 229d 63 229e 122 229u 104 229v 112 229v v 112 229w 112 229ww

.112 229x 112 229y 115a 229z 112 230 54 231 34 232 119 233 124 234 119 235 115 236

.115 243 128 244 124 248 59 249 120 250 128 251 124 252 120 253 120 254 120 255 120 256 122 257 122 258 134 258a 132 259 120 260

.120 261 120 262 120 B

Amendment No. 134 03/31/03 I

MONTICELLO NUCLEAR GENERATING PLANT RECORD OF TECHNICAL SPECIFICATION CHANGES AND LICENSE AMENDMENTS Page License

ion DPR-22

)-No.

Amend No. & Date nal 1

1/19/71 Note 2 Note 2 Note 2 2

2/20/73 Note 2 Note 2 Note 2 Note 2 Note 2 Note 2 Note 2 Note 2 3

4 6/17/74 6

8/20/74 Note 3 5

7 8

9 10 7/8/75 12 11 13 14 15 1/22/76 16 2/3/76 17 3/16/76 NOTE 3 18 4/14/76 21 5/20/76 AEC Tech Spec Chanae Issuance No. and d, Note 1 2

1/14/7 3

t 0/31J 4

12/8/7 Note 1 5 -3/2/73 1

4/28/7 6

4/3/72 7

5/4/72 8

7/2/73 9

8/24/7 10 10/2/7 11 11/27/

12 11/15/

13.3/30/7 14 5/14/7 Note 1 Note 1 Note 3 1(

15 1/15/7 16 2/3/75 17 2/26/7 18 4/10/7 Note 1 20 9/15/7 19 9/17/7 21 10/6/7 22 10/30/

NOTE 4 4/13/76 ate Major Subject Appendix A Technical Specifications incorporated in DPR-22 on 9/8/70.

Removed 5 MWt restriction 72 MOGS Technical Specification changes issued by AEC but never distributed or put into effect, superseded byTS Change 1211/15/73 72 RHR service water pump capability change

'2 Temporary surveillance test waiver Increase in U-235 allowed in fission chambers Miscellaneous Technical Specification changes,

'1 &

Respiratory Protection, & Administrative Control Changes Respiratory Protection Changes Relief Valve and CRD Scram Time Changes

'3 Fuel Densification Limits 73 Safety Valve Setpoint Change 73&

Offgas Holdup System, RWM, and 73 Miscellaneous Changes 74 8x8 Fuel Load Authorization 74 8x8 Full Power authorization Changed byproduct material allowance Changed byproduct material allowance

)/24/74 Inverted Tube (CRD) Limits 5

REMP Changes Reactor Vessel Surveillance Program Changes 5

Vacuum Breaker Test Changes 5

Corrects Errors & Provides Clarification Increased allowed quantity of U-235 5

Snubber Requirements 5

Removed byproduct material allowance 5

Suppression Pool Temperature Limits 75 Appendix K and GETAB Limits Reporting Requirements CRD Collet Failure Surveillance NSP Organization Changes Adoption of GETAB Containment Isolation Valve Testing Interim Appendix B, Section,2.4 Tech. Specs.

C Amendment No. 115 12/21/00 NSP Revis (REV Origir 1

2 3

4 5.

6 7

8 9

10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25

  • I

MONTICELLO NUCLEAR GENERATING PLANT RECORD OF TECHNICAL SPECIFICATION CHANGES AND LICENSE AMENDMENTS License DPR-22 Amend No. & Date 19 5/27/76 20 6/18/76 22 7/13/76 23 9/27/76 24 10115/76 25 10/27/76 26 4/1/77 27 5/24/77 28 6/10/77 29 9/16/77 30 9/28/77 31 10/14/77 32 12/9/77 33 1/25/78 34 A

4/14/78 35 9/i 5/78 36 10/30/78 37 11/6/78 NOTE 3 11/24/78 38 3/15/79 39 5/15/79 40 6/5/79 41 8/29/79 42 12/28/79 43 2/12/80 44 2/29/80 8/29/80 9/19/80 10/24/80 1/9/81 1/9/81 Major Subject Low Steamline Pressure Setpoint and MCPR Changes APLHGR, LHGR, MCPR Limits Correction of Errors and Environmental Reporting Standby Gas Treatment System Surveillance CRD Test Frequency Snubber Testing Changes APRS Test Method MAPLHGR Clamp at Reduced Flow Radiation Protection Supervisor Qualification REMP Changes More Restrictive MCPR Inservice Inspection Changes Reporting Requirements Fire Protection Requirements Increase in spent fuel storage capacity RPT Requirements Suppression Pool Surveillance 8x8R Authorization, MCPR Limits & SRV Setpoints Corrected Downcomer Submergence Incorporation of Physical Security Plan into License Revised LPCI Flow Capability Respiratory Protection Program Changes Fire Protection Safety Evaluation Report MAPLHGR vs. Exposure Table MCPR & MAPLHGR Changes for Cycle 8 and Extended Core Burnup ILRT Requirements Order for Modification of License-Environmental Qualification Revised Order for Modification of License-Environmental Qualification Order for Modification of License-Environmental Qualification Records Issuance of Facility Operating License (FTOL)

Order for Modification of License Concerning BWR Scram' Discharge Systems (License conditions removed per Amendment No. 11 dated 10/8/82)

D Amendment No. 115 12/21/00 NSP Revision (REN) No.

26 27 28

-29 30 31 32 33 34 35 36 37 38 39 NOTE 1 40 41 42 43 44 45.

46 47 48 49 50 NOTE 1 NOTE 1 51 52 NOTE 1 I

MONTICELLO NUCLEAR GENERATING PLANT

-RECORD OF TECHNICAL SPECIFICATION CHANGES ANDLICENSE AMENDMENTS License

o.

Amend No. & De 1/13/81 1

2/12/81 2

3/2/81 3

3/27/81 4

3/27/81 5

5/4/81 6

6/3/81 7

6/30/81 8

11/5/81 9

12/28/81 1/19/82 10 11 12 13-14 15 16 17 18 19 20 21 22 23 24 25 26 27 28 29 30 31 5/20/82 10/8/82 11/30/82 12/6/82 12/10/82 12/17/82 4/18/83 4/17/83 11/28/83 12/30/83 1/16/84 1/23/84 2/2/84 4/3/84 5/1/84 8/15/84 9/24/84 10/31/84 11/2/84 11/16/84 11/16/84 11/27/84 ate Major Subject Order for Modification Mark I Containment Revision of License Conditions Relating to Fire Protection Modifications TMI Lessons Learned & Safety -

Related Hydraulic Snubber Additions Low voltage protection, organization and miscellaneous Incorporation of Safeguards Contingency Plan and Security Force Qualification and Training Plan into License Cycle 9 - ODYN Changes, New MAPLHGR's, RPS Response time change Inservice Inspection Program Fire Protection Technical Specification Changes Mark I Containment Modifications Inservice Surveillance Requirements for Snubbers Revised Order for Modification Mark I Containment Scram Discharge Volume New Scram Discharge Volumes RPS Power Monitor Cycle 10 Recirc Piping and Coolant Leak Detection Appendix I Technical Specifications (removed App. B)

Organizational Changes Miscellaneous Changes Steam Line Temperature Switch Setpoint Radiation Protection Program SRM Count Rate Definition of Operability Miscellaneous Technical Specification Changes RPS Electrical Protection Assembly Time Delay Scram Discharge Volume Vent and Drain Valves Miscellaneous Technical Specification Changes Cycle 11 RHR Intertie Line Addition Hybrid I Control Rod Assembly ARTS Low Low Set Logic Degraded Voltage Protection Logic EI

Amendment No. 115 12/21/00 NSP Revision (REV) N NOTE 1 53 54 NOTE 1 55 56 57 58 59 NOTE 1 60 61 62 63 64 65 66 67 68

-69 70 71 72 73 74 75 76 77 78 79 80 81

MONTICELLO NUCLEAR GENERATING PLANT RECORD OF TECHNICAL SPECIFICATION CHANGES AND LICENSE AMENDMENTS License DPR-22 Amend No. & Date 32 5/28/85 33 10/7/85 34 10/8/85 35 12/3/85 36 12/23/85 37 1/22/86 38 2/12/86 39 3/13/86 40 3/18/86 41 3/24/86 42 3/27/86 43 4/8/86 44 5/27/86 45 7/1/86 46 7/1/86 47 10/22/86 48 12/1/86 49 8/26/87 50 8/26/87 51 10/16/87 52 11/18/87 53 11/19/87 54 11/25/87 55 11/25/87 56 12/11/87 57 9/23/88 58 12/13/88 59 2/16/89 60 2/28/89 61 3/29/89 62 3/31/89 63 4/18/89 64 5/10/89 65 5/30/89 66 5/30/89 Major Subject Surveillance Requirements Screen Wash/Fire Pump (Partial)

Fuel Enrichment Limits Combustible Gas Control System Vacuum Breaker Cycling NUREG-0737 Technical Specifications Environmental Technical Specifications Administrative Changes Clarification of Radiation Monitor Requirements 250 Volt Battery Jet Pump Surveillance Simmer Margin Improvement Cycle 12 Operation Miscellaneous Changes LER Reporting and Miscellaneous Changes Single Loop Operation Offgas System Trip Rod Block Monitor APRM and IRM Scram Requirements 2R Transformer Surveillance Intervals - ILRT Schedule Extension of Operating License Cycle 13 and Misc Changes Appendix J Testing ATWS - Enriched Boron Increased Boron Enrichment Physical Security Plan Miscellaneous Administrative Changes Miscellaneous Administrative Changes Fire Protection and Detection System ADS Logic and S/RV Discharge Pipe Pressure Miscellaneous Technical Specification Improvements Containment Vent and Purge Valves NUREG-0737 - Generic Letter 83-36 Reactor Vessel Level Instrumentation F

Amendment No. 115 12/21/00 NSP Revision (REV) No.

82 83 84 85 86 87 88 89 90 91 92 93 94 95 96 97 98 99 100 101 102 103 104 105 106 107 108 109 110 111 112 113 114 115 116 I

MONTICELLO NUCLEAR GENERATING PLANT RECORD OF TECHNICAL SPECIFICATION CHANGES AND LICENSE AMENDMENTS License DPR-22 Amend No. & Date 67 6/19/89 68. 7/14/89 69 9/12/89 70 9/28/89 71 10/19/89 72 11/2/89 73 5/1/90 74 6/5/90 75 10/12/90 76 12/20/90 77 2/15/91 78 3/28/91 79 4/9/91 80 8/12/91 81 4/16/92 82 7/15/92 83 8/18/92 84 1/27/93 85 6/29/93 86 7/12/93 87 4/15/94 88 6/30/94 89 8/25/94 90 9/7/94 91 9/9/94 92 9/15/94 93 7/12/95 94 10/2/95 95 4/3/96 96 4/9/96 97 9/17/96 Major Subject Extension of MAPLHGR. Exposure for One Fuel Type SRO Requirements & Organization Chart Removal Operations Committee Quorum Requirements Relocation of Cycle-Specific Thermal-Hydraulic Limits Deletion of Primary Containment Isolation Valve Table RG 1.99, Rev 2, ISI & ILRT Combined STA/LSO Position Removal of WRGM Automatic ESF Actuation Diesel Fuel Oil Storage Miscellaneous Administrative Changes Redundant and IST Testing Alarming Dosimetry SAFER/GESTR Torus Vacuum Breaker Test Switch/EDG-Fuel Oil Tank Level Surveillance Test Interval Extension - Part I Alternate Snubber Visual Inspection Intervals Revisions to Reactor Protection System Tech Specs MELLIA and Increase Core Flow Revision to Diesel Fire Pump Fuel Oil Sampling Requirements Revisions to Control Rod Drive Testing Requirements Revised Coolant Leakage Monitoring Frequency Average Planar Linear Heat Generation Rate (APLHGR)

Specification & Minimum Critical Power Ratio Bases Revisions Removal of Chlorine Detection Requirements and Changes to Control Room Ventilation System Requirements Revisions to Radiological Effluent Specifications Secondary Containment System and Standby Gas Treatment System Water Level Setpoint Change Change in Safety Relief Valves Testing Requirements Revised Core Spray Pump Flow Standby Gas Treatment and Secondary Containment Systems MSIV Combined Leakrate, and Appendix J, Option B Purge and Vent Valve Seal Replacement Interval Implementation of BRWOG Option -D core Stability Solution and re-issue of pages 11,- 12, 82 and 231 to reflect pages issued by NRC amendments.

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Amendment No. 115 12/21/00 NSP Revision (REV) No.

117 118 119 120 121 122 123 124 125 126 127 128 125 130 131 132 133 134 135 136 137 138 139 140 141 142 143 144 145 146 147

MONTICELLO NUCLEAR GENERATING PLANT RECORD OF TECHNICAL SPECIFICATION CHANGES AND LICENSE AMENDMENTS License DPR-22 Amend No. & Date 98 7/25/97 99 10/29/97 11/25/97 100 100a 101 4/20/98 4/30/98 8/28/98 102 9/16/98 103 12/23/98 104 105 106 107 108 109 110 111 12/24/98 03/19/99 10/12/99 11/24/99 12/8/99 2/16/00 8/7/00 8/18/00 112 8/18/00 113 10/2/00 114 11/30/00 115 11 5a 12/21/00 02/13/01 116 03/01/01 117 03/07/01 118 03/09/01 118a 05/10/01 Maior Subiect Bases changes on containment overpressure and number of RHR pumps required to be operable. Reissue pages 207, 209, 219, 229k, 229p, 230, 245 to reflect pages issued by NRC amendments.

SLMCPR for Cycle 18 and reissue pages vi, 155, 202, 207, 219, 229u Reissue pages a, b, g, iii, vi, 14, 25a, 155, 198y, 198z, 202, 207, 209, 219, 229k, 229p, 229r, 229u, 230, 245 SLMCPR for Cycle 19 Reissue all pages.

Reactor Coolant Equivalent Radioiodine Concentration and Control Room Habitability Monticello Power Rerate Surveillance Test Interval/Allowed Outage Time Extension Program - Part 2 Revision of Statement on Shift Length & other Misc Changes CST Low Level HPCI/RCIC Suction Transfer Revised RPV-PT Curves & remove SBLC RV setpoint Reactor Pressure Vessel Hydrostatic and Leakage Testing Testing Requirements for Control Room EFT Filters Safety Limit Minimum Critical Power Ratio for Cycle 20 Transfer of Operating Authority from NSP to NMC Transfer of Operating License from NSP to a New Utility Operating Company Emergency Filtration Train Testing Exceptions and Technical Specification Revisions Alternate Shutdown System Operability Requirements Safety/Relief Valve Bellows Leak Detection System Test Frequency Administrative Controls and Other Miscellaneous Changes Bases Change to Reflect Modification 98Q145 Installed Control Room Toxic Gas Air Supply Relocation of Inservice Inspection Requirements to a Licensee Program Reactor Water Cleanup (RWCU) System Automatic Isolation and Miscellaneous Instrumentation System Changes Revision of Standby Liquid Control System Surveillance Requirements Bases Change - 50°F Loop Temperature, Bus Transfer &

Rerate Correction H

Amendment No. 119 04/05/01 NSP Revision (REV) No.

148 149 NOTE 5 150 NOTE 6

MONTICELLO NUCLEAR GENERATING PLANT RECORD OF TECHNICAL SPECIFICATION CHANGES AND LICENSE AMENDMENTS License DPR-22 Amend No. & Date 119 04/05/01 11 9a 06/28/01 120 07/24/01 121 07/25/01 122 08/01/01 122a 10/22/01 123 123a 10/26/01 10/25/01 Major Subiect Fire Protection Technical Specification Changes Bases Change - Added information on cooldown rate Relocation of Radiological Effluent Technical Specifications to a Licensee-Controlled Program Clarify air ejector offgas activity sample point and operability requirements Relocation of Inservice Testing ReqVuirements to a Licensee-Controlled Program Bases Change - Remove scram setpoints sentence and correct typo-Control Rod Drive and Core Monitoring Technical Specification Changes Bases Change.- Change to reflect new operation of drywell to suppression chamber vacuum breaker valve position indicating lights -

124 10/30/01 Relocation of Technical Specification Administrative Controls Related to Quality Assurance Plan 1 24a 12/05/01 Bases Change - Change to reflect revised Technical Specification definition of a containment spray/cooling subsystem 125 12/06/01 Safety Limit Minimum Critical Power Ratio for Cycle 21 126 01/18/02 Elimination of Local Suppression Pool Temperature Limits 126a 02/15/02 Bases Change - Change reflects relocation of sample point for the offgas radiation monitor 127 05/31/02 Missed Surveillance Requirement Technical Specification Changes 128 06/11/02 Changes to the Technical Specifications Revised Reference Point for Reactor Vessel Level Setpoints, Simplification of Safety Limits, and Improvement to the Bases 128a 07/11/02 Bases Change - Correct Drywell to Suppression Chamber Vacuum Breaker Indicating Light Description 129 08/27/02 Revise Technical Specifications and Surveillance Requirements Relating to Standby Diesel Generators 1 29a 09/12102 Bases Change - Change to Snubber Operability Description 129b 09/12/02 Bases Change - Remove Language That Implies Trip Settings Can Be Modified By Deviation Values 130 09/23/02 Containment Systems Technical Specification Changes 130a 09/26/02 Bases Change - HPCI - Change Wording / HPCI & RCIC -

Enhance with Wording Consistent with NUREG-1 433-Rev 1 1

Amendment No. 130a 09/26/02-NSP Revision (REV) No.

MONTICELLO NUCLEAR GENERATING PLANT RECORD OF TECHNICAL SPECIFICATION CHANGES AND LICENSE AMENDMENTS -

License DPR-22 Amend No. & Date 131 10/02/02 132 133 133a 02/04/03 02/24/03 03/28/03 134 03/31/03 135 04/22/03 135a 04/24/03 Major Subiect Update the Multiplier Values for Single Loop Operation Average Planar Linear Heat Generation Rate (APLHGR)

Conversion to Option B for Containment Leak Rate Testing Revision to Pressure-Temperature Curves Bases Change - Adequate Reactor Steam Flow for HPCI/RCIC Testing One-Time Extension of Containment Integrated Leak-Rate Test Interval Boiling Water Reactor Vessel and Internals Project Reactor Pressure Vessel Integrated Surveillance Program Bases Change - Clarify description of head cooling Group 2 valves.

1. License Amendment or Order for Modification of License not affecting Technical Specifications.
2.

Technical Specification change issued prior to. 10 CFR revisions which require issuance of Technical Specification changes as License Amendments.

3.

Modification to Bases. No Technical Specification change or License Amendment issued..

4.

Technical Specification change numbers no longer assigned beginning with Amendment 15.

5.

Pages reissued 11/25/97 to conform with NRC version. Revision number of effected pages not changed..

6.

All pages reissued using iNTERLEAF in different font. Using NRC Amendment Nos. and issue date. For Bases and Table of Contents, spelling errors corrected and editorial corrections made and all Amendment Nos. changed to 1 Ooa. For remaining Tech Spec pages, no other changes made and current Amendment Nos. used.

J Amendment No. 135a 04/24/03 NSP Revision (REV) No.

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